The disposal of large quantities of toxic materials such as high level radioactive wastes stored in spent reactor fuel storage pools, or generated in the reprocessing of spent nuclear power reactor fuel, or generated in the operation and maintenance of nuclear power plants, is a problem of considerable importance to the utilization of nuclear power. It is generally accepted that the most promising approach is to convert these radioactive wastes to a dry solid form which would render such wastes chemically, thermally and radiolytically stable.
The problem of dry solid stability of radioactive wastes is closely related to the safety of human life on earth for a period of more than 20,000 years. For example, radioactive wastes usually contain the isotopes Sr.sup.90, Pu.sup.239, and Cs.sup.137 whose half lives are 28 years, 24,000 years, and 30 years, respectively. These isotopes alone pose a significant threat to life and must be put into a dry, solid form which is stable for thousands of years. The solid radioactive waste form must be able to keep the radioactive isotopes immobilized for this length of time, preferably even in the presence of an aqueous environment. The radioactive wastes are produced in high volumes and contain long-lived, intermediate-lived, and short-lived radioactive ions and some non-radioactive ions. These solutions can be highly corrosive and it is difficult, if not impractical, to reduce them to concentrated forms for further processing or storage.
The two most popular types of commercial reactors both of which produce low level wastes are the Boiling Water Reactor (B.W.R.) and the Pressurized Water Reactor (P.W.R.). In a typical Pressurized Water Reactor (P.W.R.), pressurized light water circulates through the reactor core (heat source) to an external heat sink (steam generator). In the steam generator, where primary and secondary fluids are separated by impervious surfaces to prevent contamination, heat is transferred from the pressurized primary coolant to secondary coolant water to form steam for driving turbines to generate electricity. In a typical Boiling Water Reactor (B.W.R.), light water circulates through the reactor core (heat source) where it boils to form steam that passes to an external heat sink (turbine and condenser). In both reactor types, the primary coolant from the heat sink is purified and recycled to the heat source.
The primary coolant and dissolved impurities are activated by neutron interactions. Materials enter the primary coolant through corrosion of the fuel elements, reactor vessel, piping, and equipment. Activation of these corrosion products adds radioactive nuclides to the primary coolant. Corrosion inhibitors, such as lithium, are added to the reactor water. A chemical shim, boron, is added to the primary coolant of most P.W.R.'s for reactivity control. These chemicals are activated and add radionuclides to the primary coolant. Fission products diffuse or leak from fuel elements and add nuclides to the primary coolant. Radioactive materials from all these sources are transported around the system and appear in other parts of the plant through leaks and vents as well as in the effluent streams from processes used to treat the primary coolant. Gaseous and liquid radioactive wastes (radwaste) are processed within the plant to reduce the radioactive nuclides that will be released to the atmosphere and to bodies of water under controlled and monitored conditions in accordance with federal regulations.
The principal methods or unit operations used in the treatment of liquid radwaste at nuclear power plants are filtration, ion exchange, and evaporation.
Liquid radwastes in a P.W.R. are generally segregated into five categories according to their physical and chemical properties as follows:
a. Clean Waste includes liquids which are primarily controlled releases and leaks from the primary coolant loop and associated equipment. These are liquids of low solids content which are treated in the reactor coolant treatment system. PA1 b. Dirty or Miscellaneous Waste includes liquids which are collected from the containment building, auxiliary building, and chemical laboratory; regeneration solutions from ion-exchange beds; and solutions of high electrical conductivity and high solids content from miscellaneous sources. PA1 c. Steam Generator Blowdown Waste is condensate from the steam that is removed (blowdown) periodically to prevent excessive solids buildup. PA1 d. Turbine Building Drain Waste is leakage from the secondary system that is collected in the turbine building floor sump. PA1 e. Detergent Waste includes liquids from the laundry, personnel decontamination showers, and equipment decontamination. PA1 a. High-Purity Waste includes liquids of low electrical conductivity (&lt;50 .mu.mho/cm) and low solids content, i.e., reactor coolant water that has leaked from the primary reactor system equipment, the drywell floor drain, condensate demineralizer backwash, and other sources of high-quality water. PA1 b. Low-Purity Waste includes liquids of electrical conductivity in excess of 50 .mu.mho/cm and generally less than 100 .mu.mho/cm; i.e., primarily water from floor drains. PA1 c. Chemical Waste includes solutions of caustic and sulfuric acid which are used to generate ion exchange resins as well as solutions from laboratory drains and equipment decontamination. PA1 d. Detergent Waste includes liquids from the laundry and personnel decontamination showers. PA1 (A) The dead ion exchange resin containing radioactive waste is mixed with cement and cast in forty gallon barrels. PA1 (B) The bottoms from evaporators which contain the radioactive contaminated boric acid and the solutions used to regenerate the ion exchange columns are mixed with cement powder and cast in forty gallon metal or plastic barrels. PA1 (C) The filters containing particulate forms of radioactive waste are usually encased in cement in metal or plastic barrels. PA1 (i) The thermal expansion coefficient is so low that it can only be matched by the core glass when the loading is very low (e.g. less than 5 weight % for the UK composition, see Example 25). PA1 (ii) Because of the high collapsing temperature (about 1300.degree.-1400.degree. C.) it may cause volatilization of Cs and other nuclear wastes. PA1 (1) The preform is immersed in a solution containing the dopant ions at a pH between 9 to 13.5, preferably between 10 and 13, for a time which depends on the wall thickness and the desired concentration of dopants. Typically, the immersion time is between 1 hour and 7 days. The pH of the solution is preferably adjusted with NH.sub.4 OH. For maximum speed of ion exchange, the solution is saturated with the desired dopant ions. Usually the dopants are introduced into the solution as nitrate compounds. However, chlorides and carbonates can be used.
Liquid radwastes in a B.W.R. are generally segregated into four categories according to their physical and chemical properties as follows:
The liquid radwastes from both types of reactors are highly dilute solutions of radioactive cations, and other dissolved radioactive materials as well as undissolved radioactive particles or finely divided solids.
A practical process for disposing of radioactive materials in a dry solids form having high resistance to leaching and other forms of chemical attack would not only be suitable for the disposal of radioactive nuclear wastes, but also for the fabrication of radioactive sources useful in industry, medicine, and in the laboratory.
Heretofore, there did not exist any practical, foolproof means for the safe disposal, storage and immobilization of pernicious radioactive waste material. Present day storage containers do not provide sufficient isolation and immobilization of such radioactive material, sufficient long-term resistance to chemical attack by the surroundings, and sufficient stability at high temperature.
Currently low level radioactive waste, that is radioactive waste generated at reactor sites, is disposed of in the following manner:
These cement barrels are transported to low level radioactive waste sites and buried six feet deep in the ground. At least one of the sites is in the United States Eastern States and exposed to substantial rainfall. In Europe, these barrels are buried at sea. In both cases water will first corrode the metal then the cement and will relatively quickly expose the radioactive ions for leaching into the ground water or sea water. Because the U.S. burials are only a few feet deep, the contaminated water can readily intermix with streams, lakes and rivers, thus, entering the ecosphere. The rationale for this practice is the assumption that upon sufficient dilution the radioactivity becomes harmless.
Some of the most serious nuclear wastes are cesium and strontium which are biologically similar to sodium and calcium. They have thirty year half lives indicating that they should be isolated from the ecosphere for at least three hundred years (ten half lives). At Bikini, the experts assumed that dilution had made the island inhabitable after decades in which no atomic explosions were performed, yet when the population was returned to the island its health was deleteriously effected. It has since been realized that plants and animal life biologically reconcentrate these radioactive elements back up to dangerous levels.
Thus, the "safe" concentration of radioactive waste must be much lower than accepted values and a more durable substitute for cement is needed. The present invention presents a safe alternative to the cement-solidification of low level waste.
Another route heretofore suggested is the so-called dry solids approach which involves the fixation of the waste materials in glasses via mixing with glass-forming compositions and melting to form glasses. This approach offers some improvement regarding isolation and decrease in the rate of release of radioactive elements when the outer envelopes or containers are destroyed. Further, such glasses reamin relatively more stable at high temperatures than plastic and are generally more chemically durable in saline solutions than are metals. Glasses with high chemical durability and low alkali ion conductivity suitable for this prior art technique are formed at very high temperatures, e.g., 1800.degree. C. and higher. Prior processes utilizing such high melting glass-forming compositions are economically unsound and moreover, cause a dangerous problem due to the risk of volatilization of pernicious radioactive materials. Furthermore, this prior procedure is restricted to dry solid radioactive wastes and provides no solution to the high volumes of liquid radioactive wastes produced by the operation and maintenance of nuclear reactors, by the current practice of storing spent fuels in pools of water, and by spent reactor fuel recovery systems.
In view of the overall difficulties of handling radioactive material, and especially in view of the danger of volatilization of radioactive material into the atmosphere, attention has been directed to using glass compositions having relatively low melting temperatures, that is to say, using glass compositions with SiO.sub.2 contents as low as 27 weight percent. While the problem of volatilization of radioactive materials is reduced, it is not completely controlled. Moreover, the resultant glass composition exhibits greatly reduced chemical durability and increased ion diffusion rates for the radioactive materials present therein. The greater this diffusion rate, the lower is the ability of the glass to keep the radioactive materials immobilized in its matrix. For long-term containment of radioactive waste, demanded under present day standard, these prior glass compositions are inadequate.
U.S. Pat. No. 3,640,888 teaches the production of neutron sources by encapsulating californium-252 in glass using the steps of packing an open-ended vitreous tube with a porous powder of quartz having an organic liquid ion exchange material sorbed thereon, passing an aqueous solution containing californium-252 cations through the powdered quartz, drying and heating the powdered quartz and tube in air to oxidize and volatilize the organic liquid ion exchange material resulting in the non-volatile oxide of californium-252, and then fusing the tip and powder contents to form a vitreous body containing the californium-252 oxide. The patent, however, does not disclose, teach or suggest the use of porous glass as the tube, in packing for the tube or in a stopper for the tube nor does it disclose or suggest any method or technique for concentrating and safely disposing of radioactive wastes as by burying in strong, durable, leach-resistant glass containers.
U.S. Pat. No. 1,533,794 teaches the packaging of radium emanations in a glass capillary tube followed by sealing the ends of the tube and thus enclose emanations previously introduced into the tube. There is no teaching, however, of any method for concentrating and encapsulating radwaste.
U.S. Pat. Nos. 2,336,227; 2,340,013; 2,522,524; 3,364,148 and 4,073,579 relate to the treatment of porous glass with non-radioactive ions (radioactive ions in the case of 3,364,148) followed by heating to close the pores which contain the ions. U.S. Pat. No. 3,147,225 discloses refractory particles, which contain no or minor amounts of silica and preferably are crystalline, within which particles a specifically selected radioactive cation is firmly fixed for use in self-luminous markers, liquid level indicators and other applications.
U.S. Pat. No. 3,116,131 discloses the method of binding expanded silica particles with a binder and shaping and curing into a desired form, followed by impregnation with a solid dessicant, e.g., sodium hydroxide, and followed by impregnation with a radioactive gas and steam to absorb the water vapor followed by capillary condensation thereby entraining the radioactive gas in the pores after which the pores are closed by heating.
U.S. Pat. No. 3,959,172 discloses the method of forming and reacting a mixture of silicate or other source of silicon, a radionuclide waste and a metal cation to produce complex metalosilicate crystals which entrap the radionuclide waste. U.S. Pat. Nos. 3,451,940 and 3,849,330 disclose the utilization of a thermite reaction to form a complex polysilicate product containing the radioactive wastes.
U.S. Pat. No. 3,167,504 discloses the purification of radioactive waste liquid by absorption on a synthetic zeolite which is then sealed in a suitable container for burial.
U.S. Pat. Nos. 3,114,716; 3,262,885; 3,365,578 and 4,020,004 each deal with various techniques involving the preparation of glass-forming mixtures followed by firing to form a glass.
U.S. Pat. No. 3,093,593 discloses methods for disposing of radioactive wastes by forming porous ceramic pieces from clays and other silicates followed by prefiring such pieces to destroy ion exchange capacity and thereafter impregnating the prefired pieces with radioactive liquid wastes. The pieces saturated with radioactive waste are then heated to vitrify them and render them non-absorptive.
U.S. Pat. No. 3,938,974 relates to glass, optical wave guide fibers and their production. Radioactive materials cannot be used in such fibers because they form color centers which absorb light. Not only does this patent fail to disclose the use of radioactive materials, the presence of such materials are inimical to the express objects of the patent.
There is no disclosure or suggestion in any of the above-identified patents of radwaste disposal methods involving the depositing of radwaste solids in a glass container, and heating to drive off non-radioactive volatile materials, e.g., water and decomposition products, while preventing the escape of radioactive materials and to collapse the container and seal it, thereby providing a durable, highly leach-resistant, long-term containment of radioactive solids.
As will be apparent hereinafter from the various aspects of applicants' contributions to the art, there are provided novel methods providing novel articles for the containment of pernicious and dangerous radioactive materials over extraordinarily long periods of time. Unlike melting glass containment procedures, the methods of the present invention provide for the control of radioactive materials that are prone to volatilization at high temperatures employed in the containment procedure, thereby providing for elimination of environmental hazards due to the possible escape of volatilized radioactive material in the atmosphere and avoiding the necessity of providing elaborate recapture and/or redisposal procedures and equipment.